A conventional boiling water reactor is shown in FIG. 1. Feedwater is admitted into a reactor pressure vessel (RPV) 10 via a feedwater inlet 12 and a feed-water sparger 14, which is a ring-shaped pipe having suitable apertures for circumferentially distributing the feed-water inside the RPV. The feedwater from sparger 14 flows downwardly through the downcomer annulus 16, which is an annular region between RPV 10 and core shroud 18.
Core shroud 18 is a stainless steel cylinder surrounding the nuclear fuel core 20. Core 20 is made up of a plurality of fuel bundle assemblies 22 (only two 2.times.2 arrays of which are shown in FIG. 1). Each array of fuel bundle assemblies is supported at the top by a top guide 19 and at the bottom by a core plate 21. The core top guide provides lateral support for the top of the fuel assemblies and maintains the correct fuel channel spacing to permit control rod insertion.
The water flows through downcomer annulus 16 to the core lower plenum 24. The water subsequently enters the fuel assemblies 22, wherein a boiling boundary layer is established. A mixture of water and steam enters core upper plenum 26 under shroud head 28. Vertical stand-pipes 30 atop shroud head 28 are in fluid communication with core upper plenum 26. The steam-water mixture flows through standpipes 30 and enters steam separators 32. The separated liquid water then mixes with feed-water in the mixing plenum 33, which mixture then returns to the core via the downcomer annulus. The steam passes through steam dryers 34 and enters steam dome 36. The steam is conducted from the RPV via steam outlet 38.
The BWR also includes a coolant recirculation system which provides the forced convection flow through the core necessary to attain the required power density. A portion of the water is pumped from the lower end of the downcomer annulus 16 via recirculation water outlet 42 and forced by a centrifugal recirculation pump (not shown) into jet pump assemblies 44 (only one of which is shown) via recirculation water inlets 46. The BWR has two recirculation pumps, each of which provides the driving flow for a plurality of jet pump assemblies. The jet pump assemblies are circumferentially distributed around the core shroud 18.
The core shroud 18 (see FIG. 2) comprises a shroud flange 18a for supporting the shroud head 28; a circular cylindrical upper shroud wall 18b having a top end welded to shroud flange 18a; an annular top guide support ring 18c welded to the bottom end of upper shroud wall 18b; a circular cylindrical middle shroud wall comprising three sections 18d, 18e and 18f welded in series, with a top end of section 18d being welded to top guide support ring 18c; and an annular core plate support ring 18g welded to the bottom end of middle shroud wall section 18f and to the top end of a lower shroud wall 18h. The entire shroud is supported by a shroud support 50, which is welded to the bottom of lower shroud wall 18h, and by annular shroud support plate 52, which is welded at its inner diameter to shroud support 50 and at its outer diameter to RPV 10.
In the event of a seismic disturbance, it is conceivable that the ground motion will be translated into lateral deflection relative to the reactor pressure vessel of those portions of the shroud located at elevations above shroud support plate 52. Such deflections would normally be limited by acceptably low stresses on the shroud and its weldments. However, if the shroud weld zones have failed due to stress corrosion cracking, there is the risk of misalignment and damage to the core and the control rod components, which would adversely affect control rod insertion and safe shutdown.
Stress corrosion cracking in the heat affected zone of any shroud girth seam welds diminishes the structural integrity of shroud 18, which vertically and horizontally supports core top guide 19 and shroud head 28. In particular, a cracked shroud increases the risks posed by a loss-of-coolant accident (LOCA). During a LOCA, the loss of coolant from the reactor pressure vessel produces a loss of pressure above the shroud head 28 and an increase in pressure inside the shroud, i.e., underneath the shroud head. The result is an increased lifting force on the shroud head and on the upper portions of the shroud to which the shroud head is bolted. If the core shroud has fully cracked girth welds, the lifting forces produced during a LOCA could cause the shroud to separate along the areas of cracking, producing undesirable leaking of reactor coolant.
A known repair method for vertically restraining a weakened core shroud utilizes tensioned tie rods coupled to the shroud flange and to the shroud support plate. In addition, the shroud is restrained laterally by installation of wishbone springs which, along with the tie rod, are components of the shroud repair assembly. This shroud repair apparatus is shown in FIG. 2.
Referring to FIG. 2, a shroud restraint tie rod assembly comprises a tie rod 54 having a circular cross section. A lower end of tie rod 54 is anchored in a threaded bore formed in the end of a spring arm 56a of a lower spring 56. Tie rod 54 extends from the end of spring arm 56a to a position adjacent the outer circumferential surface of the top guide support ring 18c. The upper end of tie rod 54 has a threaded portion.
The lower spring 56 is anchored to a gusset plate 58 attached to the shroud support plate 52. The lower spring 56 has a slotted end which straddles gusset plate 58 and forms a clevis hook 56c. The clevis hooks under opposite ends of a clevis pin 60 inserted through a hole machined in the gusset plate 58. Engagement of the slotted end with the gusset plate 58 maintains alignment of lower spring 56 under the action of seismic motion of the shroud, which may be oblique to the spring's radial orientation.
The tie rod 54 is supported at its top end by an upper support assembly 62 (shown in greater detail in FIG. 4) which hangs on the shroud flange 18a. A pair of notches or slots are machined in the shroud head ring 28a of shroud head 28. The notches are positioned in alignment with a pair of bolted upper support plate segments 64 of the upper support assembly when the shroud head 28 is properly seated on the top surface of shroud flange 18a. These notches facilitate coupling of the tie rod assembly to the shroud flange.
The pair of notches at each tie rod azimuthal position receive respective hook portions 64a of the upper support plates 64. As best seen in FIG. 4, each hook 64a conforms to the shape of the top surface of shroud flange 18a and the shape of the steam dam 29. The distal end of hook 64a hooks on the inner circumference of steam dam 29.
Referring to FIGS. 3 and 4, the upper support plates 64 are connected in parallel by a top support bracket 65 and a support block 66 which forms the anchor point for the top of the tie rod. Support block 66 has an unthreaded bore 68, tapered at both ends, which receives the upper end of tie rod 54. After the upper end of tie rod 54 is passed through bore 68, a threaded nut 70 is screwed onto the threaded end of tie rod 54.
The assembly comprised of support plates 64 with hooks 64a, support block 66, tie rod 54, lower spring 56, clevis pin 60 and gusset plate 58 form a vertical load path by which the shroud flange 18ais connected to the shroud support plate 52. In the tensioned state, upper support plates 64 exert a restraining force on the top surface of the shroud flange 18awhich opposes separation of the shroud at any assumed failed circumferential weld location.
Referring to FIG. 5, the upper restraint spring 72 is a double cantilever "wishbone" design, to react the lateral seismic loads without adding bending load on the top support. The end of one arm 72a of spring 72 has an axle mounting (not shown) which is rotatably mounted in an unthreaded bore formed in an upper spring bracket 74. This allows the spring to rotate relative to the upper spring bracket 74. The end of the other arm 72b has an upper contact spacer 86 rotatably mounted there
on. Upper contact spacer 86 is designed to bear against the inner surface of the reactor pressure vessel wall. The contacting face of upper contact spacer 86 must be machined prior to its installation in dependence on the width of the downcomer annulus measured along a radius which intersects upper contact spacer 86.
The upper spring bracket 74 has a pair of parallel linear projections 76 (see FIG. 5) which slide in corresponding grooves 78 (see FIG. 4), formed on opposing surfaces of upper support plates 64, during installation of the upper spring assembly. Grooves 78 are oriented at an acute angle (e.g., 5.degree.) relative to the vertical axis of the vessel inside surface. In addition, the upper spring assembly comprises a jack bolt 80 which passes through an unthreaded bore in the upper spring bracket 74. Longitudinal displacement of jack bolt 80 relative to upper spring bracket 74 is prevented by a shoulder under the head of jack bolt 80 to rotate freely relative to upper spring bracket 74. A threaded end of jack bolt 80 projects beyond the upper spring bracket and is screwed into a threaded bore 82 in the support block 66. Threaded bore 82 is disposed parallel to grooves 78 in the upper support plates 64. Thus, as the jack bolt is rotated, the upper spring bracket 74 and upper spring 72 coupled thereto translate in parallel with grooves 78 until the upper contact spacer 86 on arm 72b is wedged against the inner surface of the reactor pressure vessel wall. The upper spring assembly is installed with enough elastic preload to prevent mechanical wear of its parts due to reactor vibration.
The upper spring 72 is installed with a desired preload against the wall of vessel 10. The amount of preload is a function of the distance which jack bolt 80 travels along bore 82 in support block 66. This mounting allows simple installation and subsequent removal, if required for reactor servicing access. When the desired amount of preload has been attained, the jack bolt is locked against further rotation relative to upper spring bracket 74 by a latching mechanism (not shown).
Lateral restraint at the elevation of the core guide support ring 18g is provided by a lower spring 56 also having a double cantilever "wishbone" design. Referring back to FIG. 2, the spring arm 56a of lower spring 56 laterally supports the shroud 18 at the core plate support ring 18g, against the vessel 10, via a lower contact spacer 88 which bears against the RPV wall. The contacting face of the lower contact spacer 88 must be machined prior to its installation in dependence on the width of the downcomer annulus measured along a radius which intersects lower contact spacer 88. The top end of spring arm 56a has a threaded bore to provided the attachment for the bottom of the tie rod 54. The member 56d connecting the upper wishbone spring 56a, 56b to the clevis hook 56c offset from the line of action between the lower end of tie rod 54 and the clevis pin 60 to provide a vertical spring compliance in the load path to the tie rod.
A middle support 90 is preloaded against the vessel wall at assembly by radial interference which bends the tie rod 54, thereby providing improved resistance to vibratory excitation failure of the tie rod. The contacting faces of the middle support 90 must also be machined prior to installation independence on the annulus width at the middle support location.
In order to produce the desired lateral restraint forces, the upper and lower spring assembly must be dimensioned in precise relationship to the annulus width at the respective elevations of the upper and lower contact spacers 86 and 88. These dimensions must be measured before the shroud repair hardware is installed to ensure proper fit of the components. In particular, the upper and lower contact spacers must be machined in dependence on the measurement results. In addition, the distance separating the tie rod and the reactor pressure vessel at the elevation of the middle support must be known with precision so that the middle support can be correctly dimensioned. Since the downcomer annulus in a boiling water reactor exposed to high radiation fields and inaccessible to maintenance personnel, it is desirable that these dimensions be measured remotely by personnal stationed at a safe distance above the reactor. Thus, there is a need for a remotely operable tool capable of precisely measuring the downcomer annulus width at a predetermined elevation.